ML20057E174
| ML20057E174 | |
| Person / Time | |
|---|---|
| Issue date: | 09/30/1993 |
| From: | Chan T Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9310080029 | |
| Download: ML20057E174 (8) | |
Text
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g UNITED STATES
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4i NUCLEAR REGULATORY COMMISSION l
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wasnincrow, o.c. 2osss-oooi September 30, 1993 ORGANIZATION:
Nuclear Utilities Management Resources Council (NUMARC) i i
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SUBJECT:
SUMARY OF MEETING WITH NUMARC ON THE STAFF'S FATIGUE '
ACTION PLAN i
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On September 17, 1993, the NRC staff met with representatives of the Nuclear Utilities Management Resources Council (NUMARC) and other interested parties.
to discuss the staff's Fatigue Action Plan as related to the assessment of-fatigue adequacy of metal components in operating reactors. A list of meeting attendees is provided in Enclosure 1.
The material distributed during the meeting is provided in Enclosure 2.
The purpose of the meeting was to initiate an information exchange dialog on the way the NRC will execute the Fatigue Action Plan..how the staff plans to interface with licensees, vendors, and research organizations, how information and views of industry can be considered during the staff's development of positions and conclusions, and how NUMARC may be able.to facilitate the-staff's assessment process.
The staff stated that it would consider all relevant information and' views i
from NUMARC, individual utilities, vendors and researcher organizations and that the use of proprietary information would be appropriately discussed and evaluated. The-staff also suggested that NUMARC provide whatever relevant information they deem to be appropriate for staff consideration compatible
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with the target milestone dates of the action plan.
In response to this suggestion, NUMARC referred to presentation handouts contained in a November 6, 1990 NRC Meeting Summary,.a NUMARC letter to the NRC dated May 29, 1991, and a BWR Owners Group letter to the NRC (date unspecified) as documents which contain relevant points on various aspects of the fatigue-issue. NUMARC also identified Electric Power Research Institute Report No. EPRI TR-100252, Metal Fatigue in Operating Nuclear Power Plants" as a document that may be relevant to Item II.2 of the staff's fatigue action plan.
NUMARC informed the staff that they were in the process of forming an Ad Hoc Advisory Committee to review and evaluate information relative to the fatigue action plan, and expressed a desire to have periodic meetings with the staff to discuss such issues. The staff stated that information. exchanges are welcomed but that it was too soon to propose specific dates.
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Terence L. Chan, Chief iping and Pipe Support Section q
Mechanical Engineering Branch 9310000029 930930 Division of Engineering PDR REV9P ER C'
Office of Nuclear Reactor Regulation.
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Enclosures:
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List of Attendees g
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Meeting Handouts p g 7g y 0600 M E O5IS C@Y (I#64 #h)
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Fatigue Meeting Summary,
Distribution:
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EMEB RF TMurley/FMiraglia i
JPartlow WRussell i
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JWiggins BLiaw SNewberry PKuo Slee RParkhill i
KWichman Meeting Attendees
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DATE OFFICIAL RECORD COPY FILENAME: G:\\CHAN\\NUMARC. SUM' i
9 ENCLOSURE I SEPTEMBER 17. 1993. LIST OF ATTENDEES MEETING WITH NUMARC I
591 ORGANIZATION Terence Chan NRC/NRR 1
John Fair NRC/NRR i
Sam Lee NRC/NRR i
Debbie Jackson NRC/NRR.
l Paul Norian NRR/RES Kurt Cozens NUMARC:
J Dave Modeen NUMARC Roger Huston TVA R
T. H. Liu Westinghouse l
Bob Borsum.
B&W Nuclear Technology Lynn Connor Southern Technical Services
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John Juliano NUS Antony Pfeffer Serch Licensing / Bechtel
DKIDSURE 2 FATIGUE ACTION PLAN 1
1 DEFINITION OF ISSUES I
In developing criteria for the evaluation of applications for license renewal, the staff developed a draft branch technical position on fatigue evaluation procedures. Subsequent discussions within the staff and between the staff and i
the industry identified three major issues regarding the fatigue evaluation of candidate plants for license renewal (and current operating plants). These issues are:
l.
Many older vintage nuclear power. plants have components of the reactor coolant pressure boundary that were designed to codes that did not require the explicit fatigue analysis required by the current ASME Code.
'l A concern with the adequacy of the fatigue design of these ' components fer the plant life was identified.
2.
Current test data show that the ASME design fatigue curves may not be conservative for nuclear power plant primary system environments. A concern with the adequacy of the fatigue design of' components designed-using these ASME curves was also identified.
L 3.
The appropriate' corrective action to be taken-when the calculated fatigue allowable limits have been exceeded (CUF>1) is the subject of' l
controversy. A staff position regarding this issue is needed.
DISCUSSION OF OPEN ISSUES For older vintage plants, components of the reactor coolant pressure boundary were designed to codes, such as ANSI B31.1, that did not require an explicit l
fatigue analysis of the components.- Because the ASME Code currently requires a fatigue evaluation of the components of the reactor coolant pressure l
boundary, this leads to a question with respect to the adequacy in terms of fatigue resistance of these older vintage plants.
In order to assess the l
fatigue resistance of the older vintage plants, an actual' fatigue evaluation l
of a sample of the components in these plants is planned.- This sample will be selected using the results of fatigue analyses from similar systems or i
l components in plants for which the fatigue analyses have been performed as a guide in selecting. critical locations.
In addition, some recent test data.indicata that the effects'of the LWR environments could significantly reduce the fatigue resistance of materials.
The ASME Code design fatigue curves were based primarily on strain-controlled l
fatigue tests of small polished specimens at room temperature in air..
Although factors of safety were applied to the~best fit curves to. cover effects such as size and data scatter, some of the recent test data indicate that these factors of safety may not be. adequate to. encompass the environmental effects.
In order to assess the significance of.the recent' test data, an actual fatigue evaluation of a sample of components in plants where Code fatigue analyses have been performed is planned. These evaluations wil1~
use interim or proposed fatigue curves that account for the environmental test data. The sample will be selected based on the most critical locations, identified by the existing Code fatigue analyses. The new fatigue evaluations L
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2 will remove mservatism, where appropriate, contained in the original fatigue analyses. This evaluation is intended to determine the impact on existing plant components of a proposed revision of the Code design fatigue curves that would account for the environmental effects.
Another major issue that has evolved from the discussions relating to the environmental effects on the fatigue curves is the appropriate corrective action required when the Code fatigue allowable limits have been exceeded (CUF>1). The staff needs to develop a regulatory position on this issue.
ACTION PLAN Phase I - Short Tern Actions 1.
Osvelop a proposed staff position paper on licensee required actions for CUF>1.0. The paper will clarify the staff's position regarding exceeding the licensing basis Code criteria and_the position will only apply to those facility's where the current licensing basis includes Code required fatigue analyses.
If the staff decides to implement new i
requirements as a result of the evaluations perfomed in this action plan, then the backfit analysis discussed in Phase III Item 4 will be required.
In developing the position paper regarding required actions for a CUF>1.0, past staff actions regarding ' exceeding licensing basis Code criteria will be researched.
For example, the staff has issued-several bulletins regarding piping analysis which contained required corrective actions for cases where calculated stresses exceed Code allowable stresses.
In addition, the staff has recently issued a 9eneric position covering piping system operability determinations.
Estimated Completion Date: Oscomber 1, 1993 Estimated Level of Effort:
12 staff weeks 2.
Obtain a set of interim fatigue design curves from RES. This effort has been completed. The interim curves were published in NUREG/CR-5999.
Phase II - Long Tern Actions j
1.
Perform a survey of current plants to determine the number of operating plants that have a fatigue analysis of the vessel, primary system components and piping. Based on the results of this' survey, select representative plants from each reactor vendor that have components of the reactor coolant system that were designed without a fatigue analysis and r6iresentative plants for which similar components were designed usins
< ASME fatigue analysis. This effort will be performed by a review of the available NRC licensing documentation.
Estimated Completion Date: September 1,1993 Estimated Level of Effort: 5 staff weeks
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1 2.
Obtain a list of the critical components in terms of fatigue usage factors from the plants that have performed the ASME fatigue ana'yses.
This effort may require coordination with the reactor vendor owner's groups.
Estimated Completion Date:_ October 1, 1993 Estimated Level of Effort: 7 staff weeks 3.
Prioritize the critical components identified in Task 2 in terns of safety significance of the components.
This effort may require coordination with,the reactor vendor owner's groups.-
Estimated Completion Date: December 1, 1993 Estimated Level of Effort:
7 staff weeks 4.
Salect example reactor coolant system components from plants designed without fatigue analyses and perform an ASME Section_ III fatigue analysis on these systems. The plants will include one from each reactor vendor and the components selected will be based on the results -
of task 3..Use both the current ASME Code and the interim fatigue design curves to perform the analysis. _In addition, the fatigue. usage factors will be computed for_both a 40 and 60 year projected life. The results of the analyses from plants that currently have fatigue analyses will be used as a guide to select appropriate component examples for this analysis.
Estimated Completion Date: May 1, 1994 Estimated Level of Effort: 32 contractor professional staff weeks 2 staff weeks 5.
Select examole reactor coolant system components from plants designed using the ASME Code current fatigue curves to assess the impact of the interim fatigue curves. The plants will include one from each reactor vendor and the components selected will be based on the results of task 3.
This evaluation will include a removal, when appropriately justified, of_the conservatism in the assumptions used in the current analysis. An example of a conservative assumption may be in the heat transfer coefficient used in the original analysis. This evaluation is intended to assess the impact on the design of a change in the design fatigue curves. This evaluation will also consider both a.40 and 60 1
year projected life.
Estimated Completion Date: May 1, 1994 Estimated Level of Effort: _32 contractor professional staff weeks 2 staff weeks
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6.
Obtain the Generic Issue 78 PRA parametric study from Research. Use these results in combination with the results of tasks 3, 4 and 5 to assess the impact of the fatigue concerns. Research estimates that the studies will be complete by 12/31/93.
Estimated Completion Date: March 1, 1994 Estimated Level of Effort: 2 staff weeks Phase III - Develop Staff Position on the Adequacy of Current Fatigue Analyses 3
1.
Obtain the latest fatigue data from all sources including foreign sourcss (i.e., the Germans and the Japanese).
Since the development of fatigue data is an ongoing effort, the latest available data will be obtained prior to developing the staff position.
Estimated Completion Date: April 1, 1994 Estimated Level of Effort: 4 staff weeks 2.
Update the interim fatigue curves using-the latest available test data.
The significance of any changes between these revised curves and.the original interim curves will be assessed in terms of the results of the Phase II example analyses.
Estimated Completion Date: May 1, 1994 Estimated Level of Effort: 2 staff weeks 3.
Meet with the current industry working groups (PVRC, ASME, etc) and obtain the latest data available from'these groups. Also obtain their input regarding the results of the staff's analysis, i
Estimated Completion Date: June 1, 1994 Estimated Level of Effort: 2 staff weeks 4.
Develop a staff position using' the available input from the fatigue studies and the industry efforts. The staff position will address: (1) whether older plants for which ASME Code fatigue analyses were not required at the time of plant licensing for the reactor coolant pressure boundary should now be required to perform a fatigue assessment of the reactor coolant pressure boundary components, and (2) whether-plants with ASME Code fatigue analyses of the reactor coolant pressure boundary-should be required to reassess the reactor coolant pressure boundary components for the impact of the new data on enviornmental concerns.
This staff position will be supported by a backfit analysis using.the-results of the PRA parametric study obtained from Research, if appropriate.
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5 Estimated Completion Date: August 1, 1994 Estimated Level of Effort:
10 staff weeks OTHER CONSIDERATIONS This is a technical action plan that is necessary to determine the scope of the problem. A regulatory licensing action plan will be developed to address the implementation of the final staff position if required.
CONTACT J. Fair 504-2759 l
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